1. Field of the Invention
The present invention relates to nuclear fuel assemblies for use in a nuclear fission reactor, and, more particularly, the present invention relates to nuclear fuel assemblies with nuclear fuel rods formed with a cladding of a composite ceramic material.
2. Description of the Related Art
During operation of a nuclear fission reactor, the function of the fuel rods within a fuel assembly is to allow transmission of the heat resulting from the fission reaction inside the fuel pellets mounted within the fuel rod while separating the radioactive material from the streaming cooling fluid, which in light water reactors is water. A nuclear fuel rod typically includes a cladding tube that houses a. stack of fuel pellets formed of uranium oxide, plutonium oxide, or a mixture thereof, and end plugs that seal both the upper and lower ends of the tube. During operation, the fuel rod claddings are subjected to heat, irradiation from the fuel pellets, and a chemical reactive environment from the streaming medium.
The fuel rod cladding in light water reactors is usually manufactured from a zirconium alloy, Zirconium alloys are used in fuel rod cladding due to their good mechanical properties, low neutron cross-section, and relatively high corrosion resistance. Different types of zirconium alloys are available for different types of light water reactors.
In spite of the favorable properties of the zirconium alloys, fuel rod claddings manufactured from a zirconium alloy are affected by the environment in the reactor (heat, radiation, chemistry environment, amount of deposition, and location of deposition on fuel rods) such that the material expands differently, or in a non-uniform manner. The expansion of the zirconium alloy creates a permanent deformation of fuel rods, for example an elongation, such that the fuel rod dimensions in relation to its original dimensions change along the life of the fuel assembly and are different from fuel rod to fuel rod. The expansion of the zirconium alloy arises anisotropically, which results in an originally straight fuel assembly becoming bent during its life in a number of directions away from its original longitudinal axis.
The permanent deformation of the zirconium alloy in the fuel rod claddings is induced by heat, irradiation from the fuel rods, and by corrosion and hydrogen pick up. The corrosion is a function of the type of chemical environment, the type of deposition, and the quantity of deposition on each rod cladding at each location. The hydrogen pick up is a function of the cooling fluid chemical environment and the deposition resulting on fuel rod claddings. Hydrogen pick up is concentrated in the form of hydrides in the zirconium alloy, which, in addition to the permanent deformation, also results in a weakening of the mechanical properties of the fuel rod cladding. It is to be noted that the quantities of hydrides inside the fuel rod cladding resulting from the hydrogen pick up varies azimuthally and along the longitudinal axis of the fuel rod, creating different mechanical properties of the fuel rod cladding at every location along the fuel rod.
In light water reactors, the water is guided along the fuel rods from the bottom to the top of the reactor. Light water reactors are controlled by means of control elements, typically control blades that are displaced into and displaced out of positions between the fuel assemblies mounted in fuel channels for boiling water reactors (BWRs) and control rods that are displaced into and displaced out of the guide thimbles of the fuel assemblies for pressurized water reactors (PWRs). Due to the fuel assembly's great length, even a small inhomogeneous permanent deformation of the fuel rods may create a large bending of the fuel assembly. Any permanent deformation of a fuel assembly results in difficulties in movement of the control elements since there could be, for example in BWRs, frictional contact between the control blades and fuel channels resulting in “slow to settle” or totally inactive control blades.
The melting temperature of zirconium alloys is around 1750° C., substantially below the maximum temperatures reached in a dry core (2400° C.) during a beyond conceivable limits accident resulting in a total dry-out core (that is, no water available to cool the nuclear reactor). This condition would allow dissipation of radioactive materials in the melt resulting during a beyond conceivable limits accident inside the reactor vessel, and create higher impacts to the environment and to the cost and duration of recovery after such an accident. While the probability of such an accident is extremely small, it stresses the importance of introducing better materials than zirconium alloys for the fuel rod cladding as the ultimate barrier against the dissipation of large quantities of radioactive materials in the environment.